1. Field of the Invention
The present invention is directed to a test device for testing the physical reactions of nuclear fuel pellets and the interaction of nuclear fuel pellets with nuclear fuel rod cladding. More particularly, the present invention is directed to a test device which models the thermal conditions existing within an operating nuclear fuel reactor.
2. The Prior Art
In commercial water-cooled nuclear reactors used for central station electric power production, the fuel is based on urania which is sheathed or clad with zirconium alloys such as Zircaloy. Experience in the nuclear industry with fuel rod of Zircaloy-clad urania has indicated several causes for fuel rod failure. Most of these causes have been corrected by improvements in fuel design specifications and improvements in the manufacturing processes. There persists one class of fuel rod failures which has yet to be eliminated and which appears to be of a fundamental nature. These failures are caused by the direct interaction between the irradiated urania fuel, including its inventory of fission products, and the Zircaloy fuel sheath, or cladding. This phenomenon has been called "fuel/cladding interaction" or fuel "pellet-cladding interaction" (PCI). The incidence of such failures is closely linked to the power history of the fuel rod and to the severity and duration of power changes. Pellet-cladding interaction fuel rod failures have occurred in both Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) as well as in Canadian Deuterium Moderated Reactors (CANDU) and Steam Generating Heavy Water Reactors (SGHWR). To ameliorate this situation, reactor operational procedures have been established which minimize the incidence of fuel rod failures by PCI. While the operational procedures have been successful in reducing the incidence of fuel failures, the procedures are inconvenient to reactor operators and are costly in terms of reduced capacity factor for plant operation and thus in reduced electrical output. There is a strong incentive to provide a remedy that would eliminate the need for these operational procedures.
As part of this remedy, it is desirable therefore to test nuclear fuel pellets and the interaction of nuclear fuel pellets with the fuel rod cladding to improve the performance of the fuel, increase the life of the fuel rod, to determine the effects of severe temperature transients on fuel pellets and cladding and so forth. Many such tests cannot be desirably conducted in an operating reactor because variables cannot be satisfactorily controlled, test conditions are too severe, or adequate instrumentation cannot survive. Therefore, it is desirable to have a test device that can model conditions in the core of an operating nuclear reactor.
During operation of a nuclear reactor, the fuel pellets, typically uranium dioxide, and the fuel rod cladding interact in three ways which effect the performance of the fuel and the life of the fuel rod: thermal interaction; chemical interaction, including reaction of decay products with cladding; and radioactive reactions, i.e., radiation damage to fuel pellets and cladding.
The prior art discloses two approaches to the problem of testing and assessing fuel pellet-cladding interaction. First, destructive testing of nuclear fuel rods after they have been in service is known in the art. Such testing cannot, however, be truly experimental since control and independent manipulation of variables is not possible in an operating nuclear reactor. In addition, reaction of fuel rods to extreme conditions that might damage the reactor cannot be tested, since the most important such tests involve the effects of repeated rapid heating, and wide temperature excursions. Additionally, the long term hostile conditions within an operating nuclear reactor do not permit effective instrumentation of such tests.
Second, tests have been conducted on a laboratory basis by using a test section of fuel rod having an electrical heating element embedded along the longitudinal axis of the fuel rod test section. Such a device clearly must heat the test sample from the inside to the outside, so that the hotest portion of the test sample is along the axis of the fuel pellet and the coolest portion is the exterior side wall of the fuel rod cladding. This "inside-out" temperature profile is quite different from the more nearly uniform temperature profile produced in fuel rods of an operating nuclear reactor. Test results from this apparatus are not as reliable as desired because the test apparatus cannot model actual reactor conditions closely, particularly when examining the effects of temperature transients involving wide temperature excursions. Currently, the prior art has no ability to model the in-core temperature profile. This failing has been identified as a major problem in fuel performance testing (see, e.g., "Simulation of Nuclear Fuel Rods by Electrically Heated Rods", S. Malang, K. Rust, Nuclear Technology, Vol. 58, July 1982, pages 53-62).
Therefore, a need exists for a method and apparatus for testing pellet-cladding interactions on a small scale in a laboratory before a particular combination of fuel pellets and fuel rods is placed into service.